Method for the preparation of concentrated anion-deficient salt solutions

ABSTRACT

CONCENTRATED ANION-DEFICIENT SALT SOLUTIONS ARE PREPARED OF THE ACTINIDE OXIDES PUO2, UO2, UO3 ANDU3O8 BY DISSOLVING ONE OR MORE OXIDES IN AN AQUEOUS SOLUTION OF THORIUM NITRATE AT A CONCENTRATION OF 4 MOLAR OR GREATER AND AT A TEMPERATURE OF 60* C. OR MORE. ANION-DEFICIENT SALT SOLUTIONS OF ACTINIDE METALS SO PRODUCED ARE USEFUL AS STARTING MATERIALS FOR THE MANUFACTURE OF CERAMIC NUCLEAR FUEL PARTICLES BY THE SOL-GEL PROCESS.

United States Patent U.S. Cl. 252-301.1 R 2 Claims ABSTRACT OF THEDISCLOSURE Concentrated anion-deficient salt solutions are prepared ofthe actinide oxides PuO U0 U0 and U 0 by dissolving one or more oxidesin an aqueous solution of thorium nitrate at a concentration of 4 molaror greater and at a temperature of 60 C. or more. Anion-deficient saltsolutions of actinide metals so produced are useful as startingmaterials for the manufacture of ceramic nuclear fuel particles by thesol-gel process.

This is a division of application Ser. No. 106,922, filed Jan. 15, 1971,now abandond.

BACKGROUND OF THE INVENTION The invention relates to the preparation of-concentrated anion-deficient salt solutions.

Anion-deficient salt solutions are for instance suitable for thepreparation of solid oxide and carbide particles.

For the preparation of spherical particles of ceramic nuclear fuel ananion-deficient solution of uranylnitrate can successfully be used as astarting material.

In the prior art these solutions have been prepared according to thefollowing methods:

(1) By dissolving U0 in concentrated uranyl nitrate solutions,

(2) By the extraction of nitric acid from stoichiometric, possiblyslightly acid uranyl nitrate solutions.

These methods show, however, the following drawbacks.

For the purpose of the first method it is necessary to have at onesdisposal a U0 of such a texture that this substance easily dissolves inthe uranyl nitrate solution.

As to the second method it is observed that extraction, whereby nitricacid is withdrawn from a stoichiometric or weakly acid uranyl nitratesolution, can only be applied to dilute uranyl nitrate solutions.Moreover, a special installation is needed for this. After removal ofthe nitric acid the solution obtained has to be brought to the requireddegree of concentration, e.g. by evaporation.

The invention aims at giving improved methods for the preparation of ananion-deficient uranyl nitrate solution. Besides it appeared thatanion-deficient actinide saltsolutions could be prepared according toseveral more methods than was formerly possible.

DETAILED DESCRIPTION OF THE INVENTION According to the invention one ormore actinide oxides as PuO U0 or lower uranium oxides then U0 aredissolved in a small volume of an acid reacting liquid. The acidreacting liquid consists of a small amount of a strong acid such as asmall amount of concentrated HNO HCl or H 80 or an aqueous solution ofan actinide salt of a strong acid as for instance UO (NO or 'Ih(NOMixtures of the above-mentioned liquids can be used too.

With a small amount of liquid is meant that in case of "ice ananion-deficient uranyl nitrate solution the uranium concentration is atleast 2 molar.

It is possible to incorporate during the preparation or thereafter smallamounts of compounds of other elements in the anion-deficient actinidesalt solution in order to improve the properties of nuclear fuelmaterial prepared from this solution.

By other compounds are meant water soluble boron, yttrium, rare earthmetals and zirconium compounds.

Examples of the preparation of mixed anion-deficient actinide saltsolutions are the dissolving of PuO in uranyl nitrate solution and of U0in thorium nitrate solution.

It has surprisingly been found that anion-deficient solutions of therequired nitrate/actinide metal ratio can be obtained by causing loweroxides than U0 to react with strong nitric acid, uranyl nitratesolution, thorium nitrate solution or mixtures of these substances inthe quantities calculated on the basis of the requirements.

The use of lower uranium oxides than U0 has the advantage of bettersolubility in acid solutions than U03. The difficulty of preparing a U0,with a suitable texture namely can be avoided.

Lower uranium oxides than U0 are the compounds U 0 and U0 These oxides,along with uranyl nitrate, are the forms in which uranium is obtainableas a basic material. They are also the forms in which uranium ispreferably conveyed.

It is therefore of importance to convert these oxides in the easiestpossible manner into the solution required for the process to beemployed.

The required anion-deficient uranyl nitrate solution may becharacterized as follows:

[U] about 3 M [U] about 1.5

It is observed that this uranium concentration is higher than that ofthe saturated stoichiometric uranyl nitrate solution.

For the preparation of ceramic fissile material a solution of this kindis first mixed with ammonialiberating agent and then solidified by beingdispersed in a phase of sufliciently high temperature, non-miscible withwater. With this method it is of great importance to start With highlyconcentrated uranium solutions.

In order to make the rate of solution of the uranium oxide in nitricacid as high as possible, it is important to prepare the U 0 by heatingin an oxidizing atmosphere, such as air or oxygen, at temperaturesbetween 600 and 900 C. At these temperatures the most volatile and/orcombustible impurities are removed and the texture of the material isstill conducive to solution.

Diificultly soluble U0 is likewise converted by this thermal processinginto easily soluble U 0 Very diffieultly soluble U0 is converted into U0 by being sintered in air at 700 C. The cubic lattice of U0 is therebychanged into the orthorhombic lattice of U 0 As the molecular volume ofU 0 is greater than that of U0 since U0 is of higher density than U 0the particles are completely crumbled. The high specific surface areasof the U 0 obtained in this way has the effect that it can now bereadily dissolved in HNO The preparation of U 0 as described above isthe ideal method of utilizing waste obtained in the preparation of theceramic fissile material. For this purpose the waste may consist eitherof unsintered waste material, possibly containing organic filtermaterial, or of sintered final product composed of U0 In accordance withthe undermentioned gross equations (1) and (2), the quantities of nitricacid used can be determined by calculation.

zuoi+snnoa z uogmomgommi 211.0 bro-1N6. 1 5

211.0, llHNO;

e uozmomaorrmi 4x120 Nofl-n oz 2 The solution tests were repeated withtwo quantities of spherical particles of U0 with a and 40% enrichmentrespectively, after they had first been converted into U 0 The resultsobtained in this way are set forth below in Table A.

It was observed that by operating in a three-necked flask with a refluxcooler the nitrous vapours had reformed a quantity of HNO TABLE A Degreeof Density, enrich- U308, G. mol Ml. HNOa, M01 HNOs/ H2O, g./cm.a, Meas-Calcu- [NOal/ merit grams U308 14AM mol U30 ml. 21 C. ured lated [U] 20%770. l 0. 917 350 5. 50 500 l. 866 2. 82 2. 82 l. 76 40% 622. 3 0. 743270 5. 25 300 1. 004 2. 95 2. 97 1. 58

The invention is further elucidated below by reference 20 to a number ofexamples.

Example I deals with the preparation of an anion-deficient uranylnitrate solution by dissolving U0 powder in nitric acid.

Example 11 deals with the processing of spherical par- 25 ticles ofunsintered U0 Example III deals with the conversion of waste materialfrom spherical particles of U0 sintered at high temperatures.

Example IV relates to the dissolving of U 0 in uranyl nitrate solution.

EXAMPLE I A solution test was carried out with natural U0 powder innitric acid with the undermentioned quantities of U02 and HNOa-Weighed-out U0 11.4854 g.=42.5 mmol of U0 HNO 3X42.5:127.5 mmol of HNOdiluted with water to 100 ml. In this example U0 Was added in portionsto the hot (-80 C.) I-INO solution. On account of the fact that duringsolution in an open beaker some losses of nitric acid occurred, slightlymore nitric acid was used than was equivalent to equation (1).

The solution obtained was found to have an NO /U ratio of 1.6.

EXAMPLE II A quantity of spherical particles of U0 was heated slowly inair to 700 C. and then kept at this temperature for another four hours.The following was obtained:

249.8 g. of U 0 or 1000:415 mmol of U 0 density 1.965 g./cm. (20.6 C.).

EXAMPLE III 644.1 grams of spherical particles of U0 (sintered at 1400C. in an atmosphere containing hydrogen), were slowly heated to 750 C.and then kept for four hours at this temperature. In this way 662.5grams of U 0 were obtained, which could readily be passed into solutionaccording to the method indicated in Example II.

EXAMPLE IV In this example a quantity of 116 g. of

was dissolved in 72 ml. of water and then boiled under reflux with 13.7g. of U 0 for 2 hours.

The clear solution obtained had a 2.49 molar content of uranium and anNO /U ratio of 1.62.

What is claimed is:

1. A method for preparing a concentrated anion-deficient actinide saltsolution containing at least one actinide oxide selected from the groupconsisting of PuO U0 U0 and U 0 said method including dissolving at atemperature of at least C., said salt in an aqueous solution of thoriumnitrate having a concentration of at least 4 molar.

2.. A method for the preparation of a concentrated anion-deficientactinide nitrate solution wherein at least one member selected from thegroup consisting of uranium dioxide, uranium trioxide and U 0 isdissolved by stirring in a heated thorium nitrate solution of atemperature of at least 60 C., and of a concentration of at least 4molar, and the solution thus obtained is therafter diluted with water.

References Cited UNITED STATES PATENTS 3,401,122 9/1968 Cogliati et al.252301.1 S 3,307,772 7/1967 Fitch et al. 252--301.1 S 3,361,676 1/1968McBride et al. 252-3011 S 3,171,715 3/1965 Kleinsteuber 252'301.1 S

CARL D. QUARFORTH, Primary Examiner R. L. TATE, Assistant Examiner U.S.Cl. X.R.

